Pressurized-water nuclear reactors comprise a reactor coolant system in which coolant for cooling the core of the nuclear reactor flows and in which at least one steam generator is placed, providing heat exchange between the coolant of the nuclear reactor, or reactor coolant, and feedwater or secondary feedwater which is heated and vaporized inside the steam generator by heat exchange with the reactor coolant.
At the time of bringing the nuclear reactor into service, feedwater is withdrawn from a storage tank and introduced into the secondary part of the steam generator.
During operation of the nuclear reactor, the feedwater introduced into the secondary part of the steam generator is heated and vaporized; the steam is recovered and dried in the upper part of the steam generators, then sent to the turbine associated with the nuclear reactor.
The steam used to operate the turbine is then recovered in a condenser and the water formed in the condenser is reheated and sent back into a feedwater storage tank, from which it is withdrawn in order to feed the steam generator.
The feedwater of the steam generator is withdrawn from the storage tank by pumps and reheated and pressurized before being injected back into the secondary part of the steam generator.
After a certain time of operation of the nuclear reactor, it is necessary to carry out an operation of refuelling the core of the nuclear reactor, in the course of which the fuel assemblies are changed inside the core and some spent assemblies, whose burn-up rate has reached the limit set by the conditions for operating the nuclear reactor, are replaced.
In order to refuel the core of the nuclear reactor, it is necessary to completely shut down the reactor and to depressurize and cool the reactor coolant system, down to a temperature which makes it possible to open the vessel head in order to access the nuclear reactor core inside the vessel, from a pit into which the nuclear reactor vessel emerges.
In order to shut down and cool the nuclear reactor to a temperature allowing the inside of the vessel to be accessed, this nuclear reactor shutdown being called a cold shutdown, the nuclear reactor is firstly shut down by dropping all the control rods into the core of the reactor in order to obtain conditions characterizing hot shutdown of the nuclear reactor.
The hot shutdown of the nuclear reactor allows the temperature in the reactor coolant system to go from temperatures characteristic of the operating reactor (300° C. to 320° C.) down to a uniform temperature characterizing hot shutdown (about 296° C.); the pressure in the reactor coolant system is kept at a constant value (about 155 bar); the power released by the core, or the residual power of the reactor, is then less than 2% of the nominal power. The temperature in the secondary part of the steam generator goes from a value of about 290° C. down to an equilibrium temperature with the reactor coolant system (296° C.), the pressure remaining substantially constant and equal to 83 bar.
In order to carry out a cold shutdown of the nuclear reactor, a cooling system, called the reactor residual heat cooling system or RRA system, is used.
The RRA system can only be brought into service when the pressure and the temperature in the reactor coolant system have reached levels substantially lower than the temperature and pressure levels of a hot shutdown. Generally, the RRA system can only be brought into service after the reactor coolant system has been cooled down to 180° C. at least, the pressure in the reactor coolant system having been reduced down to at least 30 bar. Under startup conditions of the RRA system, the secondary cooling system is in temperature equilibrium with the reactor coolant system (180° C.) and the pressure is no more than about 10 bar.
In order to go down from temperature and pressure conditions of hot shutdown to the conditions allowing the RRA system to be started up, a steam generator auxiliary feedwater system, called an ASG system, is used in a conventional manner. Feedwater is withdrawn from a storage tank of the ASG system and injected into the steam generator via a normal feedwater pipe connected to a nozzle of the steam generator. Water from the ASG system is at low temperature (between about 7° C. and 50° C.) and the injection of water by means of a tap, in the feedwater pipe of the steam generator, causes thermal shock which could lead, with repetition, to thermal fatigue of the injection line and of the nozzles of the steam generator.
This is because water from the ASG system is introduced into the injection line of the steam generator by successive manually controlled jets. Use of the ASG system is generally characterized by a wear factor which is equal to the ratio of the number of unit jets of cold water injected into the generator to the maximum acceptable number of injections. The wear factor must be as small as possible, which restricts the use of the ASG system.
The ASG system is also used during re-startup of the nuclear reactor, from conditions for bringing into service the RRA system up to conditions characterizing hot shutdown, in order to carry out injections of secondary feedwater needed to keep the level in the steam generator substantially constant.
In addition to thermal fatigue of the injection line of the steam generator nozzle produced by injection of cold water, use of the ASG system has the drawback of bringing into service an emergency system of the nuclear reactor, in the context of normal operation of the nuclear reactor.